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Against the backdrop of China's ambitious three-step nuclear energy development strategy, encompassing pressurized water reactors (PWR), fast reactors (FR), and fusion reactors (FNR), the purpose of this study is to meet the pivotal need for a specialized sub-channel analysis code tailored to the unique thermal-hydraulic characteristics of liquid metal fast reactors (LMFRs). Building upon the foundational SACOS sub-channel code, the approach involves the seamless integration of LMFR-specific models. These include the wire-wrapped model, turbulent crossflow model, and liquid metal convective heat exchange model. The utilization of advanced computational techniques, such as the SIMPLE algorithm and staggered grid methodology, ensures the completion of accurate sub-channel calculations, establishing SACOS-LMR as a robust code for thermal-hydraulic safety analysis in LMFRs. Validation of the SACOS-LMR code was conducted through a sodium transient experiment involving 37-pin bundles at the Karlsruhe Institute of Technology. The results not only demonstrate a commendable alignment between computed parameters (e.g. temperature distribution, pressure drop) and experimental values but also confirm the code's precision in transient analysis for LMFRs. Applying the validated SACOS-LMR code, an in-depth thermal-hydraulic safety analysis of the European Lead-cooled Fast Reactor (ALFRED) core was conducted. The calculated results are not only reasonable but also exhibit consistency with comparable codes, affirming SACOS-LMR's applicability for LMFR core design and thermal-hydraulic analysis. In conclusion, this research represents a significant step forward in the development of LMFR technology. SACOS-LMR, with its validated capability in both steady-state and transient analysis, stands as a sophisticated and reliable sub-channel analysis tool. It not only supports LMFR core design but also contributes to the broader global pursuit of sustainable and clean energy solutions in the nuclear energy landscape.

RELAP5 is one of the most commonly used programs for reactor thermal hydraulic analysis. Due to the early developing time, RELAP5 does not have a friendly graphical user interface and interactive interface, resulting in the lack of human-computer interaction ability. In this work, the visual interactive simulation platform VITARS was developed by Java Script language based on Node.js framework. The origin RELAP5 was reformed then integrated into VITARS, and a bidirectional data communication pipe was established between RELAP5 process and VITARS process. In VITARS, a modeling module was developed, in which all the components in RELAP5 were abstracted into different pictograms and users could complete graphical modular modeling and parameter input for RELAP5 conveniently by stitching these pictograms, thus the use threshold of RELAP5 could be reduced. Due to the data communication pipe, VITARS could send control instructions to RELAP5 for pause, run, snapshot, terminate etc. The calculation results in RELAP5 could be displayed in VITARS as dynamic data meter or dynamic time-data curve in real time, and could be saved in VITARS for future viewing and extraction. In addition, a new signal logic computation system was developed in VITARS, in which more than forty kinds of signal logic computing components were built-in such as logical gate, proportional-integral, time delay integration, and flip-flop. As a data processing center for VITARS, the new signal logic computation system could read the thermal-hydraulic calculation results from RELAP5, user’s operation instruction etc., these data could be processed then distributed to RELAP5 to replace the thermal hydraulic boundary in time dependent component, adjustable equipment parameters or origin control data in the next time step, and could also be distributed to the data display system for real-time display. Therefore, with VITARS, complex proportion integration differentiation control system could be built for RELAP5, and user could adjust valve opening, pump speed and other equipment parameters in real time just like the reactor operators. Moreover, VITARS reserves the interfaces for coupling with other reactor analysis programs. VITARS also has the ability for users to customize the running interface. User could draw any reactor systems as the display background and kinds of graphical components such as dynamic data meter, warning light, reactor shut-down switch could be added. So that VITARS could also be used as a lightweight simulator of nuclear power plants. Model of Ling’ao Nuclear Power Plant was developed and a steady-state simulation was performed in VITARS. The simulation results were compared with that from origin RELAP5 program, which indicates the calculation accuracy of VITARS. VITARS could not only improves the efficiency of RELAP5, but also enhances human-computer interaction capabilities, which could be helpful for reactor safety analysis and operator training work.

This paper proposes a novel dynamic model considering manufacturing errors and eccentricity errors to analyze the dynamics of planetary gear transmission (PGT). The dynamic model is established based on the fractional-order calculus and solved by an enhanced fourth-order Lagrange interpolation polynomials method. Three numerical examples and the vibration experiments of PGT are employed for verification. The comparison results indicate that the proposed solution method has higher solution accuracy and efficient than the existing algorithms in solving fractional equations, and the relative errors of the proposed solution method in three examples are 0.32%, 0.78% and 0.16%, respectively. The proposed dynamic model of PGT has better agreement with the experimentally measured signal compared with the integer-order dynamic model, and the maximum error and average error of the characteristic frequency amplitude between the proposed dynamic model and the measured signal are 4.76% and 3.57%, respectively. The proposed method contributes to the theoretical foundation for the signal monitoring of PGT. [ABSTRACT FROM AUTHOR]

Flexible regulation of local chemistry and band gap of perovskite quantum dots (PeQDs) is crucial for exploring their new functionalities and device applications. In this work, a strategy based on the combination of femtosecond (fs) laser‐irradiation and thermal treatment to effectively manipulate chemical composition and emitting wavelength of PeQDs in amorphous glass, is reported. The engineering of ultrafast laser‐induced thermal effect enables to induce in situ nucleation/growth of dual‐phase PeQDs within an individual glass matrix. By elevating heat‐treatment (HT) temperature, I− ions are driven to surmount the diffusion barrier into the PeQDs lattice, leading to a tunable emission wavelength ranging from 613 to 647 nm. Besides, it is verified that the temperature‐dependent diffusion rate of I− ions plays a pivotal role in affecting luminescent efficiency and color of the dual‐phase glass. Finally, fs laser direct writing of multi‐color patterns is presented, which provides a flexible method to develop new encryption/decryption technology for information security and anti‐counterfeiting. [ABSTRACT FROM AUTHOR]

Halogen‐mixed CsPb(Br/I)3 perovskite quantum dots (PeQDs) embedded glass can address the issue of stability, but suffers from low photoluminescence quantum yields (PLQYs) for the hindered in situ nucleation/growth inside the robust glass network. Uncovering the exact mechanism is highly desirable to develop high‐performance CsPb(Br/I)3@glass for commercial applications, but the topic remains unexplored. Here, based on femtosecond transient (fs‐TA) absorption, temperature‐dependent PL spectra, and theoretical calculations, a comprehensive understanding on heat‐treatment (HT) temperature‐induced modification of microstructures and carrier dynamics in the CsPb(Br/I)3@glass is build. It is evidenced that high‐temperature HT will promote more I− ions diffusion from glass matrix into CsPb(Br/I)3 lattice, leading to the retarded hot carrier (HC) cooling, and improved exciton recombination. This is attributed to the synergistic effect of the reduced effect carrier mass, the weakened carrier‐phonon coupling, the inhibited Klemens channel, and the eliminated defect states. Revealing these underlying mechanisms will empower to exert precise control and optimize PLQY of CsPb(Br/I)3@glass up to near unity. [ABSTRACT FROM AUTHOR]

Numerical reactor technology describes a variety of physical phenomena in the core of a nuclear reactor through highprecision, highresolution, high confidence and high fidelity numerical simulation method, based on highperformance calculation and multiphysics coupling, reveals the internal laws of various physical phenomena in nuclear reactors from mechanism, and accurately predicts the key safety parameters of nuclear reactors in service. It plays an important role in the whole life cycle of nuclear reactor R&D, design, safety analysis, operation support and decommissioning. This paper systematically introduces the main outcome of the major national science and technology project “Research on Key Technologies of CAP1400 Numerical Reactor” of “Largescale Advanced PWR and HTGR Nuclear Power Plant”. Firstly, the high fidelity reactor physics calculation codes based on deterministic method and Monte Carlo method were developed respectively, and then the pinbypin advanced subchannel analysis code and the fuel rod analysis code based on finemesh were developed. Based on those codes, the CAP1400 numerical reactor system with neutronics/thermalhydraulics/fuel performance multiphysics coupling was established. The numerical reactor system was verified and validated against the international benchmark VERA provided by CASL and AP1000 startup experimental data. For VERA problems, the comparison with the results of MC21/CTF coupling codes shows that the relative deviation of the power distribution of numerical reactor system is less than 25% for both Monte Carlo method and deterministic method. The maximum critical boron concentration deviation is less than 30 ppm. For AP1000 reactor of Sanmen unit 2, the deviations of developed numerical reactor compared with the measurements are as follows: 1) The maximum deviations of gray rod worth are -103 pcm, 73 pcm and 57 pcm, respectively; 2) The maximum relative deviation of black rod worth are 32%, -17% and 08%, respectively; 3) The ITC deviations are -0.597 pcm/℃, 0.280 pcm/℃, and 1570 pcm/℃, respectively. Compared with the industry codes, numerical reactor codes have higher precision. After verification and validation, the numerical reactor codes were applied to simulate the startup experiments of CAP1400 reactor. Taking the Monte Carlo results as references, the absolute deviation of the deterministic numerical reactor code for the critical boron concentration under HZP ARO condition is -236 ppm. The maximum absolute deviation for the value of the gray rod group (MA to MD) is -52 pcm, and for the black rod group, the maximum relative deviation of the value of (M1, M2, AO, SD1 to SD6) is 268%. The relative deviation of the total worth of black rods is -008%. Assembly power distributions of 14 core conditions were compared between Monte Carlo code and deterministic code. The most positive value of relative deviation is 224%, the most negative value is -247%, the average value is -006% and the standard deviation is 102%. The above numerical results show that the developed numerical reactor analysis program has high calculation accuracy and can directly serve the design verification, startup and operation support of CAP1400.